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Bang, In Cheol (방인철)

Department
Department of Nuclear Engineering(원자력공학과)
Website
http://neths.unist.ac.kr/
Lab
Nuclear Thermal Hydraulics and Reactor Safety Lab. (원자력 열수력 및 안전 실험실)
Research Keywords
원자력시스템, 원자력안전, 원자력계통 설계 및 안전해석, 나노유체, 임계열유속, Nuclear System Design, Nuclear Safety, CHF, Boiling Enhancement technology, Nanofluid
Research Interests
The lab supports to fill important needs for more thorough treatments of the processes of energy (heat) generation in nuclear processes, the transport of that energy by the reactor coolant to the power cycle, and the limitations imposed by the transport mechanism on the design of nuclear reactor. The research has been focused on the development of enhanced heat transfer technologies in nuclear fission and fusion systems, and the study of fundamental thermal-hydraulic phenomena ultimately related to nuclear safety and economics. The lab is also interested in development of advanced engineered features for nuclear safety systems or safer nuclear reactor based on the innovative ideas. The works are engaged in both experimental and analytical research and development (R&D).
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Issue DateTitleAuthor(s)TypeViewAltmetrics
2018-04Thermal-hydraulic phenomena inside hybrid heat pipe-control rod for passive heat removalKim, Kyung Mo; Bang, In CheolARTICLE825 Thermal-hydraulic phenomena inside hybrid heat pipe-control rod for passive heat removal
2018-04Thermal-hydraulic analysis of a 7-pin sodium fast reactor fuel bundle with a new pattern of helical wire wrap spacerPark, Seong Dae; Seo, Han; Jeong, Yeong Shin, et alARTICLE672 Thermal-hydraulic analysis of a 7-pin sodium fast reactor fuel bundle with a new pattern of helical wire wrap spacer
2018-03Chromia coating with nanofluid deposition and sputtering for accident tolerance, CHF enhancementSon, Gyu Min; Kim, Kyung Mo; Bang, In CheolARTICLE663 Chromia coating with nanofluid deposition and sputtering for accident tolerance, CHF enhancement
2018-01Numerical analysis on spatial universality of similarity technique inside molten salt reactor systemSeo, Seok Bin; Shin, Yukyung; Bang, In CheolARTICLE990 Numerical analysis on spatial universality of similarity technique inside molten salt reactor system
2018-01Risk mitigation strategy by Passive IN-core Cooling system for advanced nuclear reactorsSeo, Seok Bin; Kim, In Guk; Kim, Kyung Mo, et alARTICLE767 Risk mitigation strategy by Passive IN-core Cooling system for advanced nuclear reactors
2018-01Analysis of hydrogen and dust explosion after vacuum vessel rupture: Preliminary safety analysis of Korean fusion demonstration reactor using MELCORMoon. Sung Bo; Lim, Soo Min; Bang, In CheolARTICLE654 Analysis of hydrogen and dust explosion after vacuum vessel rupture: Preliminary safety analysis of Korean fusion demonstration reactor using MELCOR
2017-10Flow visualization and heat transfer performance of annular thermosyphon heat pipeKim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin, et alARTICLE986 Flow visualization and heat transfer performance of annular thermosyphon heat pipe
2017-07Hydraulic control rod drive mechanism concept for passive in-core cooling system (PINCs) in fully passive advanced nuclear power plantKim, In Guk; Bang, In CheolARTICLE716 Hydraulic control rod drive mechanism concept for passive in-core cooling system (PINCs) in fully passive advanced nuclear power plant
2017-02Heat transfer characteristics and operation limit of pressurized hybrid heat pipe for small modular reactorsKim, Kyung Mo; Bang, In CheolARTICLE734 Heat transfer characteristics and operation limit of pressurized hybrid heat pipe for small modular reactors
2017-02Risk-reduction of passive decay heat removal system by using gallium-water for UCFR and SMRSeo, Seok Bin; Kim, In Guk; Bang, In CheolARTICLE930 Risk-reduction of passive decay heat removal system by using gallium-water for UCFR and SMR
2016-12Development of passive in-core cooling system for nuclear safety using hybrid heat pipeKim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk, et alARTICLE591 Development of passive in-core cooling system for nuclear safety using hybrid heat pipe
2016-10Hydrodynamic cavitation characteristics of an orifice system and its effects on CRUD-like SiC depositionKim, Seong Man; Bang, In CheolARTICLE975 Hydrodynamic cavitation characteristics of an orifice system and its effects on CRUD-like SiC deposition
2016-09Visual Study of Ex-Pin Phenomena for SFR with Metal Fuel Under Initial Phase of Severe Accidents by using SimulantsHeo, Hyo; Park, Seong Dae; Jerng, Dong Wook, et alARTICLE666 Visual Study of Ex-Pin Phenomena for SFR with Metal Fuel Under Initial Phase of Severe Accidents by using Simulants
2016-08High performance all-carbon composite transparent electrodes containing uniform carbon nanotube networksYun, Hyung Duk; Kwak, Jinsung; Kim, Se-Yang, et alARTICLE1525 High performance all-carbon composite transparent electrodes containing uniform carbon nanotube networks
2016-06Natural Circulation with DOWTHERM RP and its MARS Code Implementation for Molten Salt-Cooled ReactorsShin, Yukyung; Seo, Seok Bin; Kim, In Guk, et alARTICLE1120 Natural Circulation with DOWTHERM RP and its MARS Code Implementation for Molten Salt-Cooled Reactors
2016-06Length effect on entrainment limit of large-L/D vertical heat pipeSeo, Joseph; Bang, In Cheol; Lee, Jae-YoungARTICLE655 Length effect on entrainment limit of large-L/D vertical heat pipe
2016-04Comparison of Flooding Limit and Thermal Performance of Annular and Concentric Thermosyphons at Different Fill RatiosKim, Kyung Mo; Bang, In CheolARTICLE613 Comparison of Flooding Limit and Thermal Performance of Annular and Concentric Thermosyphons at Different Fill Ratios
2016-04Experimental Study on a Novel Liquid Metal Fin Concept Preventing Boiling Critical Heat Flux for Advanced Nuclear Power ReactorsPark, Seong Dae; Kim, Ji Hyun; Bang, In CheolARTICLE978 Experimental Study on a Novel Liquid Metal Fin Concept Preventing Boiling Critical Heat Flux for Advanced Nuclear Power Reactors
2016-03Numerical study of in-vessel retention under the gallium-water external reactor vessel cooling system using MARS-LMRKang, Sarah; Park, Seong Dae; Kim, In Guk, et alARTICLE1022 Numerical study of in-vessel retention under the gallium-water external reactor vessel cooling system using MARS-LMR
2016-03Hybrid heat pipe based passive cooling device for spent nuclear fuel dry storage caskJeong, Yeong Shin; Bang, In CheolARTICLE1194 Hybrid heat pipe based passive cooling device for spent nuclear fuel dry storage cask

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